1. Field
This invention relates generally to nuclear reactor systems, and in particular, to a method and apparatus for refueling a nuclear reactor.
2. Description of Related Art
A pressurized water reactor has a large number of elongated fuel assemblies mounted within an upright reactor vessel. Pressurized coolant is circulated through the fuel assemblies to absorb heat generated by nuclear reactions in fissionable material contained in the fuel assemblies. The primary side of such a nuclear reactor power generating system which is cooled with water under pressure comprises an enclosed circuit which is isolated from and in heat exchange relationship with a secondary circuit for the production of useful energy. The primary side comprises the reactor vessel enclosing a core a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. In conventional nuclear plants of that type each of the parts of the primary side comprising the steam generator, a pump and a system of pipes which are connected to the reactor vessel form a loop of the primary side.
For the purpose of illustration, FIG. 1 shows a simplified conventional nuclear reactor primary system, including a generally cylindrical pressure vessel 10 having a closure head 12 enclosing a nuclear core 14. A liquid coolant, such as water or borated water, is pumped into the vessel 10 by pumps 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above described loops are connected to a single reactor vessel 10 by reactor coolant piping 20.
An exemplary conventional reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertically co-extending fuel assemblies 22, for the purpose of this description, the other vessel internal structures can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals function to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in FIG. 2), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel through one or more inlet nozzles 30, flows down through an annulus between the reactor vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly to a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies are seated and through and about the fuel assemblies 22. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, a lower core support plate having the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44.
The upper internals 26 can be supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate 40 primarily by a plurality of support columns 48. Each support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plate 40.
Rectilinearly moveable control rods 28 which typically include a drive shaft or drive rod 50 and a spider assembly 52 of neutron poison rods, are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and the top of the upper core plate 40. The support column 48 arrangement assists in retarding guide tube deformation under accident conditions which could detrimentally affect control rod insertion capability.
To control the fission process, a number of control rods 28 are reciprocally moveable in guide thimbles located at predetermined positions in the fuel assemblies 22. Specifically, a control rod mechanism positioned above the top nozzle of the fuel assemblies supports a plurality of control rods. The control rod mechanism (also known as a rod cluster control assembly) has an internally threaded cylindrical hub member with a plurality of radial extending flukes or arms that form the spider 52 previously noted with regard to FIG. 2. Each arm is interconnected to a control rod 28 such that the control rod assembly mechanism 72 is operable to move the control rods 28 vertically within the guide thimbles within the fuel assemblies to thereby control the fission process in the fuel assembly 22, under the motive power of the control rod drive shaft 50 which is coupled to the control rod mechanism hub, all in a well known manner.
The upper internals 26 also have a number of in-core instrumentation that extend through axial passages within the support columns 48 and into instrumentation thimbles generally, centrally located within the fuel assemblies. The in-core instrumentation typically includes a thermocouple for measuring the coolant core exit temperature and axially disposed neutron detectors for monitoring the axial and radial profile of neutron activity within the core.
Nuclear power plants, which employ light water reactors require periodic outages for refueling of the reactor. New fuel assemblies are delivered to the plant and temporarily stored in a fuel storage building, along with used fuel assemblies which may have been previously removed from the reactor. During a refueling outage, a portion of the fuel assemblies in the reactor are moved from the reactor to the fuel storage building. A second portion of the fuel assemblies are moved from one support location in the reactor to another core support location in the reactor. New fuel assemblies are moved from the fuel storage building into the reactor to replace those fuel assemblies which were removed. These movements are done in accordance with a detailed sequence plan so that each fuel assembly is placed in a specific location in accordance with an overall refueling plan prepared by the reactor core designer. In conventional reactors, the removal of the reactor internal components necessary to access the fuel and the movement of the new and old fuel between the reactor and the spent fuel pool in the fuel storage building is performed under water to shield the plant maintenance personnel. This is accomplished by raising the water level in the refueling cavity and canal that is integral to the plant's building structure. The water level of more than 20 feet provides shielding for the movement of the reactor internal structures and the fuel assemblies.
Refueling activities are often on a critical path for returning the nuclear plant to power operation, therefore, the speed of these operations is an important economic consideration for the power plant owner. Furthermore, the plant equipment and fuel assemblies are expensive and care must be taken not to cause damage or unnecessary radiation exposure due to improper handling of the reactor components that have to be removed to access the fuel assemblies, the fuel assemblies or fuel transfer equipment. The precision of these operations is also important since the safe and economical operation of the reactor core depends upon each fuel assembly being in its proper location. A typical pressurized water reactor needs to be refueled every 18 to 24 months.
Commercial power plants employing the conventional designs generally illustrated in FIGS. 1 and 2 are typically on the order of 1,100 megawatts or more. More recently, Westinghouse Electric Company LLC has proposed a small modular reactor in the 200 megawatt class. The small modular reactor is an integral pressurized water reactor with all primary loop components located inside the reactor vessel. The reactor vessel is surrounded by a compact, high pressure containment. Due to both limited space within the containment and the low cost requirement for integral pressurized light water reactors, the overall number of auxiliary systems including those associated with refueling needs to be minimized without compromising safety or functionality. For example, the compact high pressure containment associated with the design of some small modular reactors does not allow for the incorporation of a large floodable cavity above the reactor vessel in which the transferred components can be shielded. Even in conventional designs, it would be desirable to reduce the amount of flooding required for refueling to save time and the expense of the operation.
Accordingly, it is an object of this invention to provide a method and apparatus for shielding the movement of a fuel assembly that does not require flooding of the containment.
It is a further object of this invention to provide such a method and apparatus that does not require the addition of motorized components to deploy shielding during fuel movement.
It is an additional object of this invention to provide such a method and apparatus which practically does not require additional storage space over that currently required.